US10643758B2 - Treatment method for volume reduction of spent uranium catalyst - Google Patents
Treatment method for volume reduction of spent uranium catalyst Download PDFInfo
- Publication number
- US10643758B2 US10643758B2 US15/960,166 US201815960166A US10643758B2 US 10643758 B2 US10643758 B2 US 10643758B2 US 201815960166 A US201815960166 A US 201815960166A US 10643758 B2 US10643758 B2 US 10643758B2
- Authority
- US
- United States
- Prior art keywords
- uranium
- solution
- solid
- spent
- treatment method
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Active, expires
Links
Images
Classifications
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
- G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor
- G21F9/28—Treating solids
- G21F9/30—Processing
- G21F9/308—Processing by melting the waste
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
- G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor
- G21F9/04—Treating liquids
- G21F9/06—Processing
- G21F9/12—Processing by absorption; by adsorption; by ion-exchange
-
- B—PERFORMING OPERATIONS; TRANSPORTING
- B01—PHYSICAL OR CHEMICAL PROCESSES OR APPARATUS IN GENERAL
- B01D—SEPARATION
- B01D21/00—Separation of suspended solid particles from liquids by sedimentation
-
- C—CHEMISTRY; METALLURGY
- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22B—PRODUCTION AND REFINING OF METALS; PRETREATMENT OF RAW MATERIALS
- C22B60/00—Obtaining metals of atomic number 87 or higher, i.e. radioactive metals
- C22B60/02—Obtaining thorium, uranium, or other actinides
- C22B60/0204—Obtaining thorium, uranium, or other actinides obtaining uranium
- C22B60/0217—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes
- C22B60/0221—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes by leaching
- C22B60/0226—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes by leaching using acidic solutions or liquors
- C22B60/0234—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes by leaching using acidic solutions or liquors sulfurated ion as active agent
-
- C—CHEMISTRY; METALLURGY
- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22B—PRODUCTION AND REFINING OF METALS; PRETREATMENT OF RAW MATERIALS
- C22B60/00—Obtaining metals of atomic number 87 or higher, i.e. radioactive metals
- C22B60/02—Obtaining thorium, uranium, or other actinides
- C22B60/0204—Obtaining thorium, uranium, or other actinides obtaining uranium
- C22B60/0217—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes
- C22B60/0221—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes by leaching
- C22B60/0247—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes by leaching using basic solutions or liquors
-
- C—CHEMISTRY; METALLURGY
- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22B—PRODUCTION AND REFINING OF METALS; PRETREATMENT OF RAW MATERIALS
- C22B60/00—Obtaining metals of atomic number 87 or higher, i.e. radioactive metals
- C22B60/02—Obtaining thorium, uranium, or other actinides
- C22B60/0204—Obtaining thorium, uranium, or other actinides obtaining uranium
- C22B60/0217—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes
- C22B60/0252—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes treatment or purification of solutions or of liquors or of slurries
- C22B60/0278—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes treatment or purification of solutions or of liquors or of slurries by chemical methods
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
- G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor
- G21F9/04—Treating liquids
- G21F9/06—Processing
- G21F9/10—Processing by flocculation
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
- G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor
- G21F9/28—Treating solids
- G21F9/30—Processing
- G21F9/301—Processing by fixation in stable solid media
- G21F9/302—Processing by fixation in stable solid media in an inorganic matrix
- G21F9/305—Glass or glass like matrix
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02P—CLIMATE CHANGE MITIGATION TECHNOLOGIES IN THE PRODUCTION OR PROCESSING OF GOODS
- Y02P10/00—Technologies related to metal processing
- Y02P10/20—Recycling
Definitions
- the present invention relates to a treatment method for volume reduction of a spent uranium catalyst.
- CH 2 CHCN
- the uranium used in the spent uranium catalyst is depleted uranium (uranium-234; 0.001%, uranium-235; 0.194%, uranium-238; 99.805%), and the amount is 4-9 wt %.
- the spent uranium catalyst also includes antimony at the concentration of 15-25 wt % and iron at the concentration of about 5 wt %.
- the catalyst support silica includes Si at the concentration of 50-60 wt %.
- the disposal requirement for uranium-bearing waste to a low-intermediate level radioactive waste repository operating in Korea is that “total radioactivity of alpha-emitting waste such as uranium waste should be less than 3,700 Bq/g.
- the activity is equivalent to 15.3% of natural uranium activity and to 25.2% of the activity of depleted uranium with 0.194 wt of U-235.
- the clearance activity concentration of uranium waste is derived on the basis of evaluation of dose assessment scebarios meeting two conditions of individual dose criterion of less 10 ⁇ Sv/yr, and of group dose creiterion of less than 1 man ⁇ Sv.
- the release limit for natural radionuclides proposed by International Atomic Energy Agency (IAEA RS G-1.7) is up to 1 Bq/g. This indicates that the content of depleted uranium used in the uranium catalyst to meet the activity concentration is equivalent to less than about 0.007 wt % in the solid waste.
- the radioactivity meets the acceptance criteria of the radioactive waste repository operated in Gyeongju area.
- the disposal cost is around KRW 15 million per 200 L drum, but it is expected that it will reach KRW 20 million per drum due to the increase of treatment costs and additional costs in the future.
- the volume of the final wastes is predicted to be at least 10,000 to 20,000 drums.
- the disposal cost of the low-level waste disposal site constructed in Korea is very expensive and it is constructed for the disposal of mainly nuclear power plant wastes, indicating that more effective and efficient method has to be executed in the course of treating the spent uranium catalyst generated in private sector in order to increase the usability of the waste disposal site and to minimize the secondary waste production, considering the economic efficiency of the treatment. It is very difficult to completely separate uranium components from the spent uranium catalyst. Also, the separated uranium is depleted uranium so that it is not economically efficient.
- the concentration of all alpha nuclides in the separated high-purity uranium waste is higher, which is about 14,600 Bq/g, than the limit for the disposal of all alpha nuclide wastes in domestic low-level waste repositories, which is 3,700 Bq/g. Thus, the waste is not accepted in the repositories.
- the silicon component in the support which occupies most of the volume of the spent uranium catalyst, can be selectively dissolved and discharged after being purified to meet the clearance criteria, and thus if the remaining insoluble ingredients and uranium can be treated as mixed to meet the disposal limit of radioactivity for the waste repositories and if the wastewater also generated in the course can meet the discharge criteria, the disposal volume of the spent uranium catalyst can be minimized.
- Such idea of the spent uranium catalyst treatment method above can make an important progress to increase waste treatment efficiency and disposal site usability.
- Patent Reference 1 Korean Patent No. 10-1316925 (Patent Reference 1) is a domestic patent having such technical background.
- the spent uranium catalyst is repeatedly dissolved in an acid and alkali solutions, so that the device and the procedure are complicated due to the solid-liquid separation of the insoluble materials during that and due to the repeated transfer of the solid material to the acid and alkali dissolution tanks.
- a large amount of wastewater to be treated in a subsequent process is generated due to the use of a large amount of acid and alkali solutions.
- the additional steps of precipitating the remaining uranium in the generated waste liquid to be uranium peroxide (UO 4 ) and eliminating uranium using an adsorbent to remove trace uranium remaining in the solution are required, which make the whole process complicated and produce more waste liquid, indicating the volume of waste liquid to be treated is increased.
- the uranium remained in the solution generated after the dissolution and precipitation of silicon dioxide is to be precipitated as uranium peroxide. Its concentration greatly affects the rate of uranium peroxide precipitation and the size of uranium peroxide particles formed.
- the present inventors confirmed that the method of the present invention is simpler than the conventional method for treating the spent uranium catalyst and can minimize the production of waste liquid.
- the present inventors further proposed a more efficient volume reduction process of a spent uranium catalyst and a method of immobilizing the final solid materials generated in the process, leading to the completion of a highly efficient volume reduction method of the waste to be disposed.
- the present invention provides a treatment method of a spent uranium catalyst comprising the following steps:
- the solid-liquid separation in step 2 can be performed by using a separation membrane coated with diatomite powder.
- the undissolved solid materials and the uranium phosphate precipitate are mixed, and a glassification agent is added thereto, followed by heat-treatment to immobilize the mixture in the form of a glass-ceramic composite medium. Therefore, the volume of waste proceeding to the disposal site is significantly reduced and the leaching characteristics of the immobilized waste is remarkably improved as well.
- the method of the present invention also uses the filtration method which is characterized by using a separation membrane coated with diatomite powder as a filter-aid for soil-liquid separation of slurry generated after dissolving the spent uranium catalyst in an alkali solution, indicating that the separation efficiency is greatly improved, compared with the conventional method, and the separation membrane can be effectively re-used as well.
- FIG. 1 is a process flow chart illustrating the treatment method of a spent uranium catalyst of the present invention designed for the volume reduction of the spent uranium catalyst.
- FIG. 2 is a graph illustrating the UO 4 precipitation depending on the initial Si concentration in the solution when the UO 4 precipitation method is used to treat uranium waste liquid.
- FIG. 3 is a graph illustrating the changes in the spent uranium catalyst particle before and after inducing the selective dissolution of the support component by immersing the spent uranium catalyst in an alkali solution.
- FIG. 4 is a set of photographs illustrating the state of the solid material separated depending on using diatomite or not in the course of the separation of the insoluble materials in a filter press.
- FIG. 5 is a XRD graph illustrating the silicon dioxide cake recovered in the filter press.
- FIG. 6 is a graph illustrating the concentration of uranium in the supernatant 2 hours after the uranium precipitation according to the type of phosphate and the concentration of phosphate ions.
- FIG. 7 is a graph illustrating the time-dependent concentration of uranium in the supernatant after uranium precipitation according to the type of phosphate and the concentration of phosphate ions.
- the present invention provides a treatment method of a spent uranium catalyst comprising the following steps:
- the solid-liquid separation in step 2 can be performed by using a separation membrane coated with diatomite powder.
- the silicon support component which occupies most of the volume of the spent uranium catalyst can be separated selectively; the concentration of uranium included therein can be reduced at the clearance level of itself, more specifically less than 1 Bq/g, to enable release to environment; the concentration of uranium included in the waste liquid generated in the course of the process can be reduced enough to be discharged (less than about 1 ppm); and the particulate wastes separated lastly can be controlled to meet the acceptance criteria of the disposal site (radioactive concentration of less than 3,700 Bq/g, immobilization of dispersed particles, etc.). Therefore, the volume reduction rate of the spent uranium catalyst can be satisfactorily excellent.
- the treatment method of a spent uranium catalyst shown in FIG. 1 of the present invention uses the chemical properties related to silicon and uranium. Particularly, the method uses many physical/chemical properties related to the selective dissolution of silicon dioxide, which occupies a large portion of the catalyst through the alkali dissolution of the spent uranium catalyst, the precipitation of dissolved silicon ions into silicon dioxide, the purification of silicon dioxide, the precipitation of uranium phosphate, the characteristics of fine particle solid-liquid separation, and the glass-ceramic solidification, etc.
- the uranium catalyst contains 4 to 9 wt % of depleted uranium containing about 0.2% of uranium-235, 15 to 25 wt % of antimony, and about 5 wt % of iron, etc. in a silicon dioxide support taking 50 to 60 wt % by the total volume.
- the spent uranium catalyst can contains bentonite added for treatment of used catalyst after for production of acrylonitrile, inorganic and organic compounds of various types of C-H-N components generated as side products during the use of catalyst (they look like black tar), other by-products generated from the catalytic reaction, and moisture, indicating the properties thereof are present in a very complicated state.
- the spent uranium catalyst requires pre-treatment before dissolution.
- tar compounds including the by-products generated during the production of acrylonitrile remain in the spent uranium catalyst, they prevent the alkali solution from contacting the support component upon dissolution of the support. Therefore, if the C-H-N compound and other containator are removed from the spent uranium catalyst through heat-treatment, the spent uranium catalyst can be more easily dissolved in the alkali solution. Also, the heat-treatment before dissolution can remove many tar-like materials, which makes the volume reduction effect of the waste itself remarkable. Therefore, it is preferred to perform the pre-heat-treatment of the target spent catalyst in order to remove water and tar-like mixed with the spent uranium catalyst.
- the heat-treatment temperature can be in the range of 300 to 1000° C., 400 to 950° C., 450 to 850° C., 500 to 800° C., 550 to 750° C., and 500 to 650° C.
- step 1 the carbonaceous impurities present on the surface of the catalyst are removed from the spent uranium catalyst through the heat-treatment. Then, the spent uranium catalyst is immersed in an alkali solution to selectively dissolve silicon dioxide, the support component of the catalyst.
- the spent uranium catalyst has a form in which U w Sb x M y O z is supported on a silicon dioxide support, and M is one or more materials selected from the group consisting of Fe, Al, Mo, V, and Bi; w, x, y, and z indicate the molar ratio of the elements constituting the oxide.
- Silicon dioxide reacts with NaOH and dissolves in the form of sodium silicate (Na 2 OSiO 2 ) n , called water glass, and can be presented as (2Na + .SiO 3 2 ⁇ ).
- the solubility of silicon ions is very high, such as 1 ⁇ 2 M at pH 14 or higher. Silicon ions are polymerized while proton ions are removed in the strong alkali condition, resulting in being a polymer ion having a high molecular weight.
- silicon of the catalyst can be mostly dissolved.
- Uranium, antimony, iron and aluminum which are very well dissolved in an acidic solution, are slightly dissolved in the alkaline condition as well.
- Uranium exists in the form of UO 2 2+ in the acidic condition and has a high solubility of 1 M or more. However, as the pH increases, UO 2 2+ becomes hydrolyzed and turns into UO 2 (OH) 2 , during which the solubility decreases to 10 ⁇ 8 M at pH 7. When the pH is continuously increased, UO 2 (OH) 2 is converted into anions such as UO 2 (OH) 3 ⁇ and UO 2 (OH) 4 ⁇ 2 ) and the solubility is increased slightly to approximately 10 ⁇ 3 M ( ⁇ 200 ppm) at pH 13 or higher. Under these alkali solution conditions, iron and antimony also remain as anions with solubilities of several tens of ppm.
- the alkali solution can be preferably one or more solutions selected from the group consisting of a sodium hydroxide solution, a potassium hydroxide solution, a lithium hydroxide solution, and an ammonium hydroxide solution.
- the molar concentration of the hydroxide ion of the alkali solution required to dissolve the silicon dioxide support varies depending on the ratio of the amount of the silicon dioxide to be dissolved and the molar concentration of the hydroxide ion, but is preferably about 2 to 4 M.
- the temperature of the alkali solution can be raised from 100 ⁇ 105° C. to the boiling point.
- the silicon dioxide support is not fully dissolved.
- the molar concentration of hydroxide ions in the alkali solution is more than 4 M, almost all the silicon dioxide component in the spent uranium catalyst are dissolved and accordingly the structure of uranium-antimony-iron oxide, the metal oxide structure of the catalyst component in the spent uranium catalyst, is destroyed, resulting in that those components remain as ultra-fine particles in the solution, which makes the solid-liquid separation of those insoluble particles difficult. Besides, these particles could remain in the solution until the step of precipitating silicon dioxide shown in FIG. 1 to affect the radioactivity concentration of the separated silicon hydroxide precipitate to be released to environment for clearance.
- the volume (mL) ratio of the alkali solution to the weight (g) of the spent uranium catalyst is 0.125 to 0.25 g/mL, preferably 0.150 to 0.225 g/mL, more preferably 0.175 to 0.210 g/mL, and most preferably 0.2 g/mL. If the volume ratio is less than 0.125, which means the amount of the spent uranium catalyst is too small for the alkali solution, the volume of the waste liquid is unnecessarily increased. If the volume ratio is more than 0.25 g/mL, which means the amount of the spent uranium catalyst is too much compared with the amount of the alkali solution to be used, the support component included in the spent uranium catalyst is not properly dissolved.
- step 2 solid-liquid separation is performed to separate the undissolved spent catalyst particles from the sodium silicate solution (Na 2 OSiO 2 ) n , the catalyst support-dissolved solution, after the dissolution in step 1.
- a solid-liquid separation device of pressurized media filtration type is used in which a solution is passed through the filter by pressurizing the device from outside.
- a filter press or a candle type filter system using a polymer membrane or a fibrous separator is selected.
- the undissolved materials contain a large amount of ultrafine particles of less than 1 ⁇ m in size.
- a separation membrane is pre-coated by putting a particulate filter-aid having a large number of pores into the filter system in advance, and then the object solution is injected to enable effective solid-liquid separation.
- the main component of the filter aid is silicon dioxide.
- the diatomite (SiO 2 : 90% or more) having different sizes of pores is used to coat the separation membrane in advance.
- the filter-aid used herein is only used for the purpose of effective solid-liquid separation, it causes another problem of increasing the volume of the final solid materials produced in the volume reduction process of the spent uranium catalyst. Therefore, the diatomite remaining in the solid of step 2 is made into a glass-ceramic medium by melting the diatomite and other materials during the immobilization step of the solid to be disposed, which is the last step of the present invention.
- a glass flux agent as Na 2 O or B 2 O 3
- the melted vitreous component can also bind to other insoluble metal oxides, so that the entire solid becomes the sable solid form during the sintering, which leads to a high volume reduction. That is, the filter-aid (diatomite) used for the solid-liquid separation of the slurry generated in step 1 not only enhances the solid-liquid separation efficiency, but also serves as a vitreous material for immobilizing the undissolved solids generated in the subsequent process, which eventually enabling additional significant volume reduction of the final solid waste.
- the mean diameter of the diatomite is not limited, but can be in the range of 30 to 40 ⁇ m.
- the uranium dissolved together with the silicon dioxide support of the spent uranium catalyst in the alkali condition is in the form of anions such as UO 2 (OH) 4 2 ⁇ , UO 2 (OH) 4 2 ⁇ , but the solubility thereof is decreased when the pH is lowered to precipitate silicon ions.
- uranium ions can be coprecipitated in the silicon dioxide precipitate as the form of UO 2 (OH) 2 by the difference in the uranium solubility at the pH during the formation of the silicon dioxide precipitate.
- a purification process comprising repeated redissolution-reprecipitation is required.
- uranium ions are in the form of anions.
- UO 2 (O 2 )(CO 3 ) 2 4 ⁇ The uranium ion species with the highest solubility is UO 2 (O 2 )(CO 3 ) 2 4 ⁇ .
- UO 2 2+ ions meet carbonate (Na 2 CO 3 ) and hydrogen peroxide (H 2 O 2 ), they form UO 2 (O 2 )(CO 3 ) 2 4 ⁇ according to the following reaction formula and the solubility thereof is as high as at least 1 M.
- the surface of the silicon dioxide particles generated in such condition has negative Zeta potential, indicating it has repulsive force to the UO 2 (O 2 )(CO 3 ) 2 4 ⁇ ions remaining in the solution, resulting in the suppression of adsorption of uranium ions. Therefore, when carbonate and hydrogen peroxide are added into the separated sodium silicate solution and pH of the solution is adjusted to 9 ⁇ 10 to precipitate silicon dioxide, uranium ions remain in the solution as much as possible and only silicon ions can be selectively precipitated in the form of silicon dioxide.
- an acid to control the pH can be nitric acid, sulfuric acid, or phosphoric acid, but sulfuric acid is preferably used in order to avoid environmental hazard of the waste solution discharged in the final process, that is, to meet the discharge limit of total nitrogen and total phosphoric acid.
- step 4 in order to perform the solid-liquid separation to separate the silicon dioxide solid formed in step 3 and the solution containing uranium ions, a solid-liquid separation device such as a filter press or a candle type filter system using media filtration is used.
- the silicon dioxide solid formed in step 3 can be easily turned into a solid cake simply by de-watering process such as the solid-liquid separation.
- the particulate filter-aid like the one used in step 2 is not required in this step.
- the silicon dioxide solid cake separated by the solid-liquid separation device such as a filter press can contain uranium solution on the surface and within its pores. Thus, to wash out the remaining uranium solution, the washing solutions such as water and an acid solution are circulated in-situ into the separated solid cake in the solid-liquid separation device.
- the separated cake is placed in 1 ⁇ 2 M NaOH, followed by heating at 60 ⁇ 80° C. to dissolve the cake completely. Then, by repeating the precipitation process of silicon dioxide used in steps 3, the purified silicon dioxide solid cake in which the uranium concentration is lowered under the self disposal level can be made.
- step 5 in order to treat the uranium wastewater generated in step 3, all the generated solutions are put into one tank, wherein uranium ions are precipitated into uranium phosphate form and the separated solution is discharged into the environment.
- the phosphate added to precipitate uranium ions into uranium phosphate can be added at the concentration of 0.1 to 6 mM in resulting solution separated through the solid-liquid separation, and preferably at the concentration of 0.2 to 4 mM, more preferably 0.3 to 2 mM, and most preferably at the concentration of 1 mM. If the concentration of the phosphate is less than 0.1 mM, the amount of the phosphate added is too small to induce the precipitation into uranium phosphate easily.
- the concentration of the phosphate added thereto is more than 6 mM, the amount of the phosphate exceeds the required amount for the possible precipitation of uranium ions and thus the concentration of phosphorus in the discharged water increases, resulting in additional step of phosphate ate removal in the solution.
- the solution from which uranium was eluted was mixed with the solution that washed the silicon dioxide solid.
- the sample solution contained uranium ions at the concentration of about 400 ppm to 1,200 ppm and silicon ions at the concentration of about 2,000 ppm to 4,000 ppm according to the mixing ratio of the solutions above.
- Approximately 0.1 M of hydrogen peroxide was added to the final uranium solution, and the pH was adjusted to 3, followed by stirring at about 200 rpm. Then, the uranium concentration changes of the solution were analyzed by ICP-OES (Analytikjena PQ9000 Elite).
- Uranium ions can form uranium phosphate displaying a very low solubility with phosphate ions (PO 4 3- ) according to the following reaction formula.
- step 6 the uranium phosphate precipitate formed in step 5 is separated from the supernatant.
- a solid-liquid separation device such as a filter press or a candle type filter system that uses media filtration without the use of a particulate filter-aid is used because the target precipitate particles are sufficiently large and uniform. If the uranium concentration in the separated supernatant is low enough, it can be discharged to environment.
- the separated uranium phosphate precipitate is combined with the undissolved solid separated in step 2, followed by solidification for disposal at the radioactive waste disposal site.
- the concentration of uranium partially dissolved with silicon dioxide in the dissolution process of the spent uranium catalyst in step 1 is very small, ranging from 200 to 300 ppm.
- This uranium is finally precipitated in step 5, so the volume of the uranium phosphate separated by precipitation is much smaller than the volume of the undissolved solid separated in step 2.
- Step 7 is a step of solidifying the solids generated in steps 2 and 6 for final disposal.
- the solids generated in steps 2 and 6 are in the form of fine particles and cannot meet the acceptance criteria of the domestic disposal site. That is, such dispersed wastes must be immobilized, and the dispersed wastes have to be prepared as solids that can satisfy certain physical properties to be treated.
- the composition is controlled and sintered to form a more sable structure in the form of glass-ceramic. In the process, the volume of the glass-ceramicized material becomes smaller than the volume of the original particulate solid, which indicates the additional volume reduction effect is obtained and accordingly the volume reduction yield of the entire waste to be treated can be improved.
- cement solidification is performed for the immobilization of the material to be treated.
- the volume of the final target material is greatly increased due to the addition of cement.
- the main component of the cement solidification medium, CaO turns into highly alkaline calcium hydroxide (Ca(OH) 2 ) when it reacts with water.
- uranium oxide can be decomposed into uranium hydroxide ions (UO 2 (OH) x y ⁇ ), resulting in the problem of uranium leaching of the solid. Due to such cement solidification characteristics, when the uranium oxide containing solid is immobilized by the cement solidification, the volume thereof is increased and undesirable solid leaching is induced.
- the water-glass solution component generated in the course of the dissolution process of step 1 remains therein.
- the water-glass (Na 2 OSiO 2 ) n solution is heated and dried, it is transformed into a solidified binder material through the dehydration process, and it can serve to immobilize the particulate solid material to be subjected to step 7.
- the solid separated in step 2 contains about 10 vol % of SiO 2 component from the undissolved support and diatomite used as a filter-aid.
- the glassification agent glass flux
- the inventive process stabilizes the waste to be treated and improves the volume reduction efficiency of the spent uranium catalyst by introducing them into the glass-ceramic solidification process.
- the glassification agent can be added at the concentration of 5 to 20 weight %, preferably 7 to 15 weight %, more preferably 9 to 13 weight %, and most preferably 10 weight % by the total silicon weight of mixture of the undissolved solid separated in step 2 and the uranium phosphate precipitate separated in step 6. If the glassification agent is added at the concentration of less than 5 weight % by the total silicon oxide weight of the waste to be immobilized, the glass-ceramic solidification process cannot be completely induced. If the glassification agent is added at the concentration of more than 20 weight % by the total silicon oxide weight of the waste to be immobilized, the vitrification can be interrupted.
- the temperature for sintering of above prepared material is 700 to 1200° C., preferably 800 to 1150° C., more preferably 900 to 1130° C. or 1000 to 1110° C., and most preferably 1100° C. If the temperature is lower than 700° C., the heat is not enough to induce glassification fully so that the volume reduction efficiency is lowered. If the temperature is higher than 1200° C., energy is wasted because of excessively high temperature more than necessary for the immobilization and of high costs for the heat treatment equipment.
- a general soda-lime glass is composed of approximately 73 wt % SiO 2 +16 wt % Na 2 O+10 wt % CaO+1 wt % Al 2 O 3 .
- a general borosilicate glass is composed of 81 wt % SiO 2 +13 wt % B 2 O 3 +4 wt % Na 2 O+10 wt % CaO+2 wt % Al 2 O 3 .
- SiO 2 has an amorphous network structure.
- the melting point, physical properties, color and various characteristics of glass can be changed by adding a glass former which forms an amorphous structure and various oxide type additives such as Na 2 O, B 2 O 3 , CaO, MgO, PbO 2 , and F 2 O 3 .
- Na 2 O plays a role as a glass solvent (Flux) to lower the melting point of the glass structure material
- B 2 O 3 acts as a glass forming agent but also plays a role as a flux to change the phase and physical properties of a ceramic material (metal oxide) by reacting to the ceramic material.
- the vitrified material can encapsulate a refractory material of metal oxides and can immobilize the entire target waste.
- This kind of structure is the glass-ceramic composite structure wherein vitreous and ceramics coexist.
- a method to reduce the volume of the entire waste wherein the water-glass component remaining in the undissolved solid left in the spent uranium catalyst-dissolution solution obtained in step 1 by using NaOH, and the filter-aid (SiO 2 diatomite) used for the solid-liquid separation of the undissolved minute solid obtained in step 2 were mixed with such additives as Na 2 O and B 2 O 3 , followed by sintering them to prepare the glass-ceramic composite material.
- the mixture material was melted and thus combined with the solids (Sb 2 O 5 , Fe 2 O 3 , and U 3 O 5 ), by which the entire target waste to be disposed became stable and more solidified, leading to the volume reduction in the end.
- the wastes thus produced have high mechanical strength and have low leaching properties of uranium from the immobilized uranium oxide.
- the treatment method of a spent uranium catalyst of the present invention comprising the steps 1 to 7 described above can be carried out in the following steps 1 to 7 when the steps are further specified:
- Step 1 Selectively Dissolving the Support Component by Immersing the Spent Uranium Catalyst in an Alkali Solution
- the spent uranium catalyst without tar-like component was added into 4 M NaOH solution, followed by dissolving at a boiling temperature (approximately 105° C.) for 4 hours with sufficient stirring. At this time, the volume ratio of the spent catalyst to NaOH was adjusted to 0.2 g (catalyst weight)/mL (NaOH solution volume).
- Table 1 shows the results of EDS analyzing the solid components before and after dissolution of the spent uranium catalyst in the alkali solution.
- the silicon content was changed from 43.8 wt % to 1.8 wt % after the dissolution. So, approximately 96% of the silicon component of the initial spent uranium catalyst was dissolved. However, antimony, uranium, and iron were dissolved only a little, unlike silicon, the composition ratio of those components in the undissolved solid remaining after the dissolution was significantly increased. Therefore, it was confirmed that only the silicon component, the support of the spent uranium catalyst, was selectively dissolved in the alkali solution.
- the volume reduction efficiency by the dissolution in step 1 was calculated by measuring the tap density values of the sample before the dissolution and the undissolved solid after the dissolution and measuring the weight thereof before and after the dissolution.
- the tap density values of the sample before the dissolution and the undissolved solid after the dissolution were 1.06 and 2.1, respectively, and the volume reduction efficiency by the dissolution was about 69%.
- pre-heat-treatment was performed at 600° C. for about 2 hours to eliminate C-H-N components from the surface of the catalyst, which is necessary step to keep these materials from interrupting the dissolution process.
- Step 2 Separating the Dissolution Solution and the Undissolved Solid Materials in the Step 1 by Solid-Liquid Separation
- step 2 of FIG. 1 after the dissolution in step 1, the undissolved solid was separated by solid-liquid separation and the silicon ions present in the dissolution solution were selectively precipitated.
- FIG. 3 shows the results of measurement of the changes in the spent uranium catalyst particle before and after the dissolution in step 1 using a particle size analyzer (Microtrac S3000).
- the particle size before the dissolution was 52.7 ⁇ m at average and the regular distribution was observed.
- the average size of the particles after the dissolution was 2.8 ⁇ m and the size distribution was irregular and the ultrafine particles in the size of less than 1 ⁇ m were observed.
- the separation membrane surface of the filter press was coated with a porous filter-aid having many fine pores and then solid-liquid separation of the slurry solution generated in step 1 was performed.
- the filter-aid used herein is diatomite (KD 801V) having an average particle size of 34 ⁇ m, which was coated on the surface of the separation membrane at the thickness of 1 to 2 mm.
- the diatomite used in this invention contains more than 90 wt % of SiO 2 , 3.8 wt % of Al 2 O 3 , and 1.3 wt % of Fe 2 O 3 , indicating that diatomite has the components that can form a glass, as described hereinbefore. So, if a glass flux such as Na 2 O or B 2 O 3 is added thereto and heated, it is vitrified and combined with the undissolved solids, resulting in the volume reduction effect along with the improvement of strength in addition to the inhibition of uranium leaching.
- a glass flux such as Na 2 O or B 2 O 3
- the separation membrane is pre-coated by injecting a particulate filter-aid having a large number of pores in the filter press system and then a target solution is injected thereto in order to perform the efficient solid-liquid separation.
- a target solution is injected thereto in order to perform the efficient solid-liquid separation.
- diatomite SiO 2 component: at least 90%
- the target ultra-fine particles are not directly contact with the separation membrane and instead the solution flows through the micropores of the filter-aid, resulting in the efficient solid-liquid separation of the ultra-fine particle mixed slurry.
- FIG. 4 is a set of photographs illustrating the state of the solid material separated depending on using diatomite or not in the course of the separation of the insoluble materials in laboratory scale and bench scale filter presses.
- the separation membrane was blocked by the fine particle sludge, the solid-liquid separation rate became very slow, and the separated solid did not form a rigid cake and instead remained as slurry. Therefore, it was confirmed that the separation membrane became clogged and contaminated with the slurry and became very difficult to be reused.
- the filter-aid when the filter-aid was used, the filter-aid first contacted with the separation membrane and the solution passed easily through the micropores in the filter-aid particles. Therefore, the solid-liquid separation rate was relatively fast, and the separated solid formed a rigid cake. Besides, after the cake separation, the surface of the separation membrane was less contaminated, indicating the separation membrane could be repeatedly reused.
- Step 3 and Step 4 Selectively Precipitating the Silicon Ions Included in the Dissolution Solution Separated in Step 2 as Silicon Dioxide, and Separating the Silicon Dioxide Generated in Step 3 by Solid-Liquid Separation and its Purification
- step 4 for the solid-liquid separation of the sludge solution solid-liquid separation was performed using a filter press equipped with a fibrous separation membrane (250 mm ⁇ 250 mm) made of polypropylene capable of filtering 0.5 to 1 ⁇ m particles, by which the hard silicon dioxide was separated into a cake form in the cell of the filter press.
- a fibrous separation membrane 250 mm ⁇ 250 mm
- the precipitated silicon dioxide has an amorphous crystal structure due to the binding form of nano-size SiO 2 particles.
- 0.5 M sulfuric acid was circulated inside the filter press, followed by circulation of distilled water.
- FIG. 5 shows the results of XRD (Bruker, D 2 Thaser) analysis of the silicon dioxide cake recovered from the filter press. As a result, it was confirmed that the cake had the same amorphous structure as the one of the commercial nanoparticle silicon dioxide, unlike the silicon dioxide peak showing the general crystallinity.
- Step 3 and step 4 were repeated to give the purified silicon dioxide cake.
- the uranium radioactivity of the silicon dioxide cake (the first one) obtained by the solid-liquid separation and the silicon dioxide purified by step 3 and step 4 was measured by using ⁇ -spectrometer (Alpha Analyst, CANBERRA).
- the radioactivity of the silicon dioxide cake primarily separated by precipitation was 3.448 Bq/g and the radioactivity of the silicon dioxide purified once was approximately 0.565 Bq/g, which was lower than the clearance criteria. Therefore, it was confirmed that the support component, silicon dioxide, selectively separated from the spent uranium catalyst was easily separated to an acceptable level which was lower than the clearance criteria.
- Step 5 Precipitating Uranium Ions as Uranium Phosphate by Adding Phosphate to the Separated Residual Solution
- uranium phosphate precipitation experiment was performed with the uranium wastewaster having the composition of Table 2 in order to confirm the uranium removal characteristics according to the phosphate type and the phosphate solution concentration, and the results are shown in FIG. 6 .
- NaH 2 PO 4 , or KH 2 PO 4 , or NH 4 H 2 PO 4 was added thereto at different concentrations, followed by stirring at 180 rpm.
- the pH was adjusted to 5. 2 hours later, the precipitation supernatant was filtered with 0.2 ⁇ m filter.
- Uranium therein was analyzed by using ICP-OES (Analytikjena PQ9000 Elite).
- the residual uranium concentration after the precipitation was reduced.
- the efficiency was most excellent.
- the residual uranium concentration was 0.016 ppm which was the ICP-OES analysis limit.
- NH 4 H 2 PO 4 was added thereto at the concentration of 1 mM or less, high efficiency was guaranteed but when NH 4 H 2 PO 4 was added at the concentration of more than 1 mM, the efficiency was decreased lower than that of when using NaH 2 PO 4 or KH 2 PO 4 . Therefore, it was confirmed that the use of NaH 2 PO 4 or KH 2 PO 4 was effective because the residual ammonium (NH 4 + ) concentration after the precipitation reaction was subject to the discharge regulation.
- FIG. 7 shows the time-dependent uranium concentration changes in the solution by the addition of NaH 2 PO 4 and KH 2 PO 4 at the concentration of 0.2 mM and 0.4 mM, respectively.
- Phosphate ions might be remaining in the supernatant after the precipitation in the process above.
- the domestic discharge limit of the phosphorus component is 2 ppm.
- the remaining phosphate ions can be easily precipitated and removed by adding ion ions (Fe 3+ ).
- Step 6 Separating Precipitate of Uranium Phosphate and the Supernatant Generated in Step 3 by Solid-Liquid Separation
- Solid-liquid separation was performed by using a filter press equipped with a fibrous separation membrane (250 mm ⁇ 250 mm) made of polypropylene capable of filtering particles in the size of 0.5 to 1 ⁇ m.
- Step 7 Mixing the Undissolved Solid Materials Separated in Step 2 with the Precipitate of Uranium Phosphate Separated in Step 6 and then Adding a Glassification Agent Thereto, followeded by Heat-Treatment to Fix the Mixture in the Form of a Glass-Ceramic Composite Medium
- the undissolved solid separated by the solid-liquid separation in step 2 and the uranium phosphate precipitate separated in step 6 were basically particulate materials, indicating immobilization was required for final disposal.
- the solid material obtained in step 2 contained a portion of the undissolved SiO 2 support, mostly undissolved Sb 2 O 5 , Fe 2 O 3 , and U 3 O 8 , and SiO 2 , the filter-aid used for the solid-liquid separation.
- step 2 and step 6 were mixed together, to which Na 2 O or B 2 O 3 , the glassification agent, was added at the ratio of about 1:7 to SiO 2 .
- This mixture was put into a mold having a diameter of 13 mm and pressed at 20 MPa to prepare a green body, which was sintered at 1,100° C. for 4 hours.
- a sintered body was prepared by compression molding by the same manner as described above except that no glassification agent was added thereto.
- the compressive strength, sintered volume reduction, final volume reduction and uranium leaching rate of the final sintered body were evaluated and the results are shown in Table 3 below.
- the leaching rate was calculated using the PCT (Product Consistency Test: ASTM-C1285) evaluation method.
- the prepared sintered body was pulverized and the particles passing through a 200-mesh filter were added to water at the ratio of 1 g:10 mL, followed by leaching at 90° C. for 7 days. Then, the uranium concentration in the solution was analyzed and calculated by the following mathematical formula.
- Table 3 presents the physical properties of the solid to be disposed.
- the compressive strength of the prepared sintered body exceeded 3.44 MPa (500 psia), which is the compressive strength required for solidified waste to be disposal in South Korea.
- a glassification agent glass flux
- B 2 O 3 was added thereto, the highest compressive strength was observed.
- the volume reduction yield after the sintering of the green body was increased as the glassification agent was added, and at this time, when B 2 O 3 was added the volume reduction rate was the best.
- the final volume reduction yield calculated from the tap density of the spent uranium catalyst and the apparent density of the sintered body was more than 60%.
- the yield was increased to 80%.
- a glassification agent was added, the undissolved uranium oxide was combined and encapsulated with the vitrified materials, resulting in the reduction of the leaching rate (improvement).
- the leaching rate herein, 1.0 ⁇ 10 ⁇ 3 g/m 2 ⁇ day, is the value of about leaching index 10 or more above leaching index 7, which is the Korean domestic disposal facility waste acceptance criteria evaluated according to ANS 16.1 method
- the volume reduction treatment method of a spent uranium catalyst of the present invention accomplished by step 1 through step 7 above can reduce the volume of the spent uranium catalyst, the treatment target, by 60 to 80% by making it in the form of a stabilized solid and the silicon dioxide and effluent generated in the course can be discharged without any problem since they satisfy the clearance criteria and discharge standard.
- the method of the present invention is simplified and efficient, compared with the conventional method, in treating a spent uranium catalyst.
Abstract
Description
-
- selectively dissolving the support component by immersing the spent uranium catalyst in an alkali solution (step 1);
- separating the dissolution solution and the undissolved solid materials in the
step 1 by solid-liquid separation (step 2); - selectively precipitating the silicon ions included in the dissolution solution as silicon dioxide (step 3);
- separating the silicon dioxide generated in step 3 by solid-liquid separation and its purification (step 4);
- precipitating uranium ions as uranium phosphate by adding phosphate to the residual solution separated in the step 4 (step 5);
- separating the uranium phosphate generated in
step 5 by solid-liquid separation (step 6); - mixing the undissolved solid materials separated in
step 2 with the precipitate of uranium phosphate separated instep 6 and then adding a glassification agent thereto, followed by heat-treatment to fix the mixture in the form of a glass-ceramic composite medium (step 7).
-
- selectively dissolving the support component by immersing the spent uranium catalyst in an alkali solution (step 1);
- separating the dissolution solution of
step 1 and the undissolved solid materials by solid-liquid separation (step 2); - selectively precipitating the silicon ions included in the dissolution solution as silicon dioxide (step 3);
- separating the silicon dioxide generated in step 3 by solid-liquid separation and its purification (step 4);
- precipitating uranium ions as uranium phosphate by adding phosphate to the residual solution separated in the step 4 (step 5);
- separating the uranium phosphate generated in
step 5 by solid-liquid separation (step 6); - mixing the undissolved solid materials separated in
step 2 with the precipitate of uranium phosphate separated instep 6 and then adding a glassification agent thereto, followed by heat-treatment to fix the mixture in the form of a glass-ceramic composite medium (step 7).
Na2SiO3+2H+=SiO2+2Na++H2O
UO2 2++H2O2+2CO3 2−=UO2(O2)(CO3)2 4−+2 H+log K=7.9
NaUO2PO4(H2O)x=Na++UO2 2++PO4 3− +xH2O Log Ksp=−24.21
KUO2PO4(H2O)x=K++UO2 2++PO4 3− +xH2O Log Ksp=−25.5
NH4UO2PO4(H2O)x=NH4 ++UO2 2++PO4 3− +xH2O Log Ksp=−26.5
UO2(O2)+2H+=UO2 2++H2O2 Log Ksp=−2.88
-
- dissolving the spent uranium catalyst by immersing thereof in an alkali solution (step 1);
- performing solid-liquid separation using diatomite powder as a filter-aid to effectively separate the dissolution solution and the undissolved microparticulate material of step a (step 2);
- making the partially co-dissolved uranium in the dissolution solution separated in
step 2 into complex ions and then precipitating the dissolved silicon ions selectively as the form of a silicon dioxide solid slurry by regulating pH (step 3); - performing solid-liquid separation to separate the uranium complex solution and the silicon dioxide solid from the solid slurry of step 3 and washing the separated solid by using acid and water (step 4);
- adding phosphate to the uranium wastewater generated in
step 4 to selectively precipitate uranium ions into uranium phosphate (step 5); - performing solid-liquid separation to separate the supernatant and the uranium phosphate precipitate generated in step 5 (step 6); and
- mixing the solids separated in
steps
TABLE 1 | |||
TK1 | Before dissolution | After dissolution | |
Si | 43.8 | 1.8 | |
Sb | 34.5 | 61.9 | |
Fe | 18.4 | 24.7 | |
U | 3.3 | 11.6 | |
Total | 100.0 | 100.0 | |
TABLE 2 | ||
ppm | mM | |
Si | 225.0 | 8.036 | |
Sb | 106.1 | 0.870 | |
K | 22.7 | 0.582 | |
U | 91.7 | 0.385 | |
Ca | 7.0 | 0.175 | |
Al | 1.80 | 0.067 | |
Mg | 1.33 | 0.055 | |
Fe | 2.14 | 0.038 | |
Zn | 2.06 | 0.032 | |
Mo | 1.00 | 0.010 | |
-
- LR (Leaching rate (g/m2 day)=Ci/(fi·SA·t)
- Ci: concentration of uranium ions measured in the solution (g/L);
- fi: uranium fraction in the sample before leaching; and
- SA: specific surface area of the sample (herein 29.9 m2/L)
TABLE 3 | ||||
Volume | Volume | |||
reduction | reduction | |||
Compressive | by heat- | from the | ||
strength | treatment | initial waste | Leaching rate | |
(MPa) | (%) | (%) | (g/m2 · day) | |
Without glass | 14.31 | 11.22 | 67.17 | 1.418 × 10−3 |
flux | ||||
Na2O | 29.31 | 19.89 | 60.15 | 1.085 × 10−3 |
B2O3 | 65.28 | 50.05 | 81.33 | 1.063 × 10−3 |
Claims (21)
Applications Claiming Priority (2)
Application Number | Priority Date | Filing Date | Title |
---|---|---|---|
KR10-2017-0089622 | 2017-07-14 | ||
KR1020170089622A KR101989910B1 (en) | 2017-07-14 | 2017-07-14 | Volume reduction treatment method of spent uranium catalyst |
Publications (2)
Publication Number | Publication Date |
---|---|
US20190019590A1 US20190019590A1 (en) | 2019-01-17 |
US10643758B2 true US10643758B2 (en) | 2020-05-05 |
Family
ID=64999585
Family Applications (1)
Application Number | Title | Priority Date | Filing Date |
---|---|---|---|
US15/960,166 Active 2038-08-04 US10643758B2 (en) | 2017-07-14 | 2018-04-23 | Treatment method for volume reduction of spent uranium catalyst |
Country Status (3)
Country | Link |
---|---|
US (1) | US10643758B2 (en) |
JP (1) | JP6586487B2 (en) |
KR (1) | KR101989910B1 (en) |
Families Citing this family (2)
Publication number | Priority date | Publication date | Assignee | Title |
---|---|---|---|---|
KR102467909B1 (en) | 2021-05-17 | 2022-11-16 | 김경덕 | Method for seperating uranium from spent uranium catalyst |
KR102641921B1 (en) | 2023-08-28 | 2024-02-28 | 엔이티 주식회사 | The batch solidification processing system for disposal of dispersibility radioactive waste |
Citations (3)
Publication number | Priority date | Publication date | Assignee | Title |
---|---|---|---|---|
JPS62191094A (en) | 1986-02-18 | 1987-08-21 | Mitsubishi Nuclear Fuel Co Ltd | Treatment of waste water containing uranium |
KR20070111891A (en) | 2006-05-19 | 2007-11-22 | 용 석 장 | The nozzle and manufacturing method thereof pressing out type machine for manufacturing noodles |
KR101316925B1 (en) | 2012-10-08 | 2013-10-18 | 한국수력원자력 주식회사 | Treatment method of spent uranium catalyst |
Family Cites Families (5)
Publication number | Priority date | Publication date | Assignee | Title |
---|---|---|---|---|
JPS60636B2 (en) * | 1979-12-25 | 1985-01-09 | 三菱マテリアル株式会社 | Treatment method for radioactive waste liquid |
KR100587157B1 (en) * | 2004-07-13 | 2006-06-08 | (주)알엔테크 | Method of Disposal of the Wasted Catalyst including Depleted Uranium |
KR100926462B1 (en) * | 2007-11-05 | 2009-11-13 | 한국원자력연구원 | Volume Reduction and Vitrification Treatment Method for Spent Uranium Catalyst Waste |
JP5544111B2 (en) * | 2009-04-16 | 2014-07-09 | 三菱原子燃料株式会社 | Method for separating, recovering and treating radioactive solid waste mainly composed of silica component |
JP2011085566A (en) * | 2009-10-14 | 2011-04-28 | 3R Corp | Method for facilitating recovery of uranium from catalyst containing uranium-antimony complex oxide |
-
2017
- 2017-07-14 KR KR1020170089622A patent/KR101989910B1/en active IP Right Grant
-
2018
- 2018-04-23 US US15/960,166 patent/US10643758B2/en active Active
- 2018-05-30 JP JP2018103459A patent/JP6586487B2/en active Active
Patent Citations (4)
Publication number | Priority date | Publication date | Assignee | Title |
---|---|---|---|---|
JPS62191094A (en) | 1986-02-18 | 1987-08-21 | Mitsubishi Nuclear Fuel Co Ltd | Treatment of waste water containing uranium |
KR20070111891A (en) | 2006-05-19 | 2007-11-22 | 용 석 장 | The nozzle and manufacturing method thereof pressing out type machine for manufacturing noodles |
KR101316925B1 (en) | 2012-10-08 | 2013-10-18 | 한국수력원자력 주식회사 | Treatment method of spent uranium catalyst |
US9181605B2 (en) * | 2012-10-08 | 2015-11-10 | Korea Atomic Energy Research Institute | Treatment method of spent uranium catalyst |
Non-Patent Citations (1)
Title |
---|
Korea Office Action, dated Dec. 27, 2018, in Korean Patent Application No. 10-2017-0089622, a related application, 6 ppl. (in Korean language). |
Also Published As
Publication number | Publication date |
---|---|
JP2019020388A (en) | 2019-02-07 |
JP6586487B2 (en) | 2019-10-02 |
US20190019590A1 (en) | 2019-01-17 |
KR101989910B1 (en) | 2019-06-18 |
KR20190008476A (en) | 2019-01-24 |
Similar Documents
Publication | Publication Date | Title |
---|---|---|
JP5734807B2 (en) | Method for treating radioactive cesium and radioactive strontium-containing substances | |
JP5722968B2 (en) | Treatment method of waste uranium catalyst | |
KR101743263B1 (en) | Treatment method of radioactive uranium waste | |
JPS6046394B2 (en) | Method for solidifying high-level radioactive waste liquid using glass | |
Tan et al. | Hydrothermal removal of Sr2+ in aqueous solution via formation of Sr-substituted hydroxyapatite | |
US10643758B2 (en) | Treatment method for volume reduction of spent uranium catalyst | |
US9480965B2 (en) | Method for preparing granulated inorganic adsorbent for radionuclides | |
Dakroury et al. | Sorption and separation performance of certain natural radionuclides of environmental interest using silica/olive pomace nanocomposites | |
EP2695167A1 (en) | Method for decontaminating radionuclides from neutron-irradiated carbon and/or graphite materials | |
Gomaa et al. | Green extraction of uranium (238U) from natural radioactive resources | |
Ouassel et al. | Application of response surface methodology for uranium (VI) adsorption using hydroxyapatite prepared from eggshells waste material: study of influencing factors and mechanism | |
KR101764865B1 (en) | Method of treating spent uranium catalyst | |
JP2013160666A (en) | Method for safely disposing burned ash containing radioactive cesium | |
Chaerun et al. | Retention mechanism of cesium in chabazite embedded into metakaolin-based alkali activated materials | |
JPS62286545A (en) | Production of ion exchange substance | |
Xian et al. | High Retention Immobilization of Iodine in B–Bi–Zn Oxide Glass Using Bi2O3 as a Stabilizer under a N2 Atmosphere | |
Kim et al. | Volume reduction of uranium catalyst waste for final disposal | |
Kim et al. | Characteristics of Cs pollucite synthesized at various Cs loadings for immobilization of radioactive Cs | |
Garino et al. | Development of iodine waste forms using low-temperature sintering glass | |
US20220401913A1 (en) | Methods of use and manufacture of silver-doped, nano-porous hydroxyapatite | |
KR101578623B1 (en) | A method of making low-melting temperature glass to immobilize radioactive cesium spent filter | |
KR101401789B1 (en) | Ceramic ingot of spent filter trapped radioactive Cesium and a method of making the same | |
Han et al. | Removal of RE3+, Cs+, Sr2+, Ba2+ from molten salt electrolyte by precipitation and solidification of glass-ceramics | |
Mimura et al. | Selective separation and recovery of cesium by ammonium tungstophosphate-alginate microcapsules | |
JP2013088184A (en) | Alkali metal iron oxide for radioactive ion adsorption/desorption and radioactive ion adsorption/desorption apparatus |
Legal Events
Date | Code | Title | Description |
---|---|---|---|
FEPP | Fee payment procedure |
Free format text: ENTITY STATUS SET TO UNDISCOUNTED (ORIGINAL EVENT CODE: BIG.); ENTITY STATUS OF PATENT OWNER: SMALL ENTITY |
|
AS | Assignment |
Owner name: KOREA ATOMIC ENERGY RESEARCH INSTITUTE, KOREA, REPUBLIC OF Free format text: ASSIGNMENT OF ASSIGNORS INTEREST;ASSIGNORS:KIM, KWANG-WOOK;LEE, KEUN-YOUNG;AHN, BYUNG-GIL;AND OTHERS;REEL/FRAME:045649/0409 Effective date: 20180201 Owner name: KOREA ATOMIC ENERGY RESEARCH INSTITUTE, KOREA, REP Free format text: ASSIGNMENT OF ASSIGNORS INTEREST;ASSIGNORS:KIM, KWANG-WOOK;LEE, KEUN-YOUNG;AHN, BYUNG-GIL;AND OTHERS;REEL/FRAME:045649/0409 Effective date: 20180201 |
|
FEPP | Fee payment procedure |
Free format text: ENTITY STATUS SET TO SMALL (ORIGINAL EVENT CODE: SMAL); ENTITY STATUS OF PATENT OWNER: SMALL ENTITY |
|
STPP | Information on status: patent application and granting procedure in general |
Free format text: DOCKETED NEW CASE - READY FOR EXAMINATION |
|
STPP | Information on status: patent application and granting procedure in general |
Free format text: NON FINAL ACTION MAILED |
|
STPP | Information on status: patent application and granting procedure in general |
Free format text: RESPONSE TO NON-FINAL OFFICE ACTION ENTERED AND FORWARDED TO EXAMINER |
|
STCF | Information on status: patent grant |
Free format text: PATENTED CASE |
|
MAFP | Maintenance fee payment |
Free format text: PAYMENT OF MAINTENANCE FEE, 4TH YR, SMALL ENTITY (ORIGINAL EVENT CODE: M2551); ENTITY STATUS OF PATENT OWNER: SMALL ENTITY Year of fee payment: 4 |